ML20006E577

Rev 1 to Sequoyah Units 1 & 2 Spent Fuel Storage Rack Criticality Analysis.

ML20006E577
Person / Time
Site:	Sequoyah
Issue date:	02/05/1990
From:	Keys T, Martin Z TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20006E575	List: ML20006E574 ML20006E577
References
PCD-D21, PCD-D21-R01, PCD-D21-R1, NUDOCS 9002260049
Download: ML20006E577 (30)

Text

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SEQUOYAH UNIT 1 AND UNIT 2  ;

,, SPENT FUEL STORAGE RACK  ;

CRITICALITY ANALYSIS l

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Document Number PCD-021 REVISION I  ;

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FEBRUARY 5,1990 .*

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' 7.A. KEYS Z. I. MARTIN TENNESSEE VAll.'EY AUTHORITY NUCLEAR FUEL DEPARTMENT 9

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. i Abstract c,- ,, . - . .

An analysis has been completed to justify an increase in the enrichment limit for the '

Sequoyah spent fuel storage rack from 4.0 wt% 235U to 5.0 wt% 235U. Since Sequoyah unit 1 and '

- both units.gnit 2 have a common spent fuel storage rack, this analysis is applicable to l

Fuel enriched to greater than 4.0 wt% with burnup less than 6,750 MWD /MTU shall be -

placed only in Sequoyah spent fuel storage rack locations with face adjacent cells which contain: ,

Fuel assemblies with accumulated burnup of at least 20,000 MWD /MTU, or

• Water. ,

The spent fuel pool water shall contain at least 2000 ppm of boron during fuel movement '

and until the configuration of the fuel in the spent fuel storage racks has been verified to -

correctly implement the criticality loading criteria. '

This will ensure that criticality safety margins are maintained. t h

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t Table of Contents 1.0 D ES CR I PTI O N O F A N A LYS IS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1. . .

' 2.02.1h1ETilODS pENCHMARKING

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2.2 COMPUTER CODE APPLICATION .............................. 4 2.3 FUEL ASSEMBLY .......................................... 5 2.4 STO R AGE RACK . . . . . . . . . . . . . . . . . . ........................ 6 2.5 ANALYTICAL MODELS ......................................6 3.0 ' ANALYSIS RESULTS .................. 8 3.1 R E S U LT S . . . . . . . . . . . . . . . . . . . . . . . . ........................

. . . . . . . . . . . . . . . . . . . . . . . . . 8-3.2 SENSITIVITY STUDIES ....................................... 8 3.3 REACTIVITY vs ENRICHMENT ........................ 11 3.4 C R E DIT FO R B U R N U P . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..........

. . . . . . . . 11 3.5 SPECIAL CASES AND POSTULATED ACCIDENTS ................. 11 3.6

SUMMARY

OF CONSERVATISMS ............................f. 13 4.0 - IMPLEMENTATION SCllEDULE AND REQUIREMENTS . . . . . . . . . . . . 15

-4.1 ~ SCHEDULE

.............................................. 15 4.2 - TECHNICAL SPECIFICATION CHANGES ........................ 15 4.3 FSAR CHANGES .......................................<...

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4.4 PROCEDURE MODIFICATIONS ......................... 16 4.5 PERSO N N EL TRAINING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... ...... 16 4.6 SPENT FUEL POOL BORATION MONITORING ,.................... 16 .

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SUMMARY

AND CONCLUSIONS ...............................

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6.0 REFERENCES

...............................:............... 19 7.0 ILLUSTRATIONS ............................................ 21 Table of Contents-

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List of Illustrations Figure 1. CASMO K-infinity Versus Burnup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 Figure 2. CASMO K-infinity Versus Enrichment . . . . . . . . . . . . . . . . . . . . . . . . . . 22 -

Figure 3. Results of KENO Benchmarking ...............................-23 Figure 4 P Figure 5.4 aR Spent Fuel Storage Rack Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 Final K-effective Determination .............................. 25 Figure ' 6. Reactivity Versus Water Density . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 f

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1.0 DESCRIPTION

OF ANALYSIS  !

DACKGROUND.

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t l The Sequoyah Nuclear Plant currently is limited to using Westinghouse Standard 17x17 l fuel assemblies' enriched to less than 4.0 wt% 235U. Future Sequoyah reloads will con- _!

sist of Westinghouse VANTAGE SH fuel assembliest with enrichments greater than 4.0 I wtS The increased enrichment is needed in order to use smaller fresh reload batch  !

fractions which increases the batch average discharge burnup from 38,000 MWD /MTU to l 48,000 MWD /MTU. .

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INTRODUCTION i

- The purpose of this criticality analysis is to justify an increase in the enrichment limit for i the Sequoyah Nuclear Plant spent fuel storage pool containing the par (Programmed "

and Remote Systems Corporation) high density spent fuel storage racks. The storage

. rack maximum fuel enrichment limit will be increased from the current 4.0 wi% 235U to I

-5,0 wt% 235U to allow the use of higher discharge burnup fuel. Higher discharge burnups j are achieved by using smaller fresh batch fractions with higher enrichments. Th,e change ,

r to a higher enrichment and discharge burnup fuel also incorporates a new fuel assembly design (VANTAGE SH) which has been evaluated to ensure that criticality safety margins  !

can be maintained. KENO and the SCALE 8 system have been used to demonstrate that  ;

the effective multiplication factor (kerr ) of the stored fuel array is ,10.95, including s i method biases, a bias for going outside the enrichment range of the KENO benchmark ~ '

I critical experiments, uncertainties due to mechanical tolerance., and the statistical un- 4 certainty associated with KENO analyses. ,

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-Using a radially infinite array of fresh fuel assemblies, the enrichment of the spent fuel racks can not be increased above 4.0 wtS Therefore, this analysis takes credit for the reactivity decrease due to burnup of the stored fuel. CASMOU.8 has been used to gen- J

( - erate equivalent enrichmentsr to account for burnup for use in KENO. An explanation of

) the equivalent enrichment concept follows.

O EQUIVALENT ENRICHMENT CONCEPT '

Reactivity.in a fuel assembly decreases as burnup accumulates due to the depletion of '

235U and the creation of fission product neutron poisons. Figure 1 displays the reactivity DESCRIPTION OF ANALYSIS I'

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decrease with burnup of an array of Westinghouse 17x17 VANTAGE SH fuel assemblies with an inillal enrichment of 5.0 wt% in the Sequoyah spent fuel racks.

Reactivity in a fuel assembly increases as the initial enrichment increases. Figure 2 displays the reactivity increase with enrichment at,zero burnup for the same configura-tion described in the previous paragraph. There exists an enrichment at zero burnup that has reactivity equivalent to the 5.0 wt% assembly with accumulated burnup. This zero '

burnup enrichment is the equivalent enrichment to the high enrichment with accumulated burnup.

KENO does not have the capability to deplete fuel assemblies. In order to model burnup, CASMO was used to determine the equivalent enrichment at zero burnup of a 5.0 wt%

fuel assembly with accumulated burnup. KENO was then used to model a radially infinite array of high enriched fresh fuel assemblies loaded in a checkerboard configuration with the low reactivity equivalent enrichment fuel assemblies determined from the CASMO analysis.

Eventually, the Sequoyah spent fuel storage racks will contain both the Standard 17x17 and the VANTAGE 5H fuel assembly designs. The VANTAGE SH design contains a slightly smaller guide tube outer diameter than that of the Standard 17x17 fuel assembly.

The VANTAGE SH assemblies also have zircaloy spacer grids as opposed to the inconel spacer grids found on the Standard 17x17 fuel assemblies. For a given enrichment, the VANTAGE SH fuel assembly has a higher reactivity than a Standard 17x17 fuel assembly.

Therefore, the limiting criticality analyses used the VANTAGE SH design since it resulted in higher reactivity values.

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i DESCRIPTION OF ANALYSIS 2

2.0 METHODS This analysis uses CASMO, a two-dimensional integral transport theory code for burnup and reactivity calculations, and KENO, a three-dimensional Monte Carlo transport theory code for rgactivity calculations. The CASMO cross sections are based on data from the ENDF/B Iveross section library with microscopic cross sections tabulated in 40 energy groups. The KENO analyses use a 27 energy group library based on ENDF/B IV cross sections.

' 2.1 BENCHMARKING 1

in order to verify the method applicabillly to criticality analyses, a series of calculations were performed using each of the methods. The results of these calculations establishes the biases and uncern Mies associated with the computer codes.

KENO I To provide a benchmark for the SCALE calculational method that is used to determine the final system reactivity, a set of 35 critical experiments were modeled to deter'mine the e methodology and cross section bias and unceriainty in predicting kerr. The criticals are representative of the Sequoyah spent fuel pool configuration, H/U ratios, enrichments,

' and spent fuel storage pool materials as indicated below: '

The critical experiments descdbed in BAW-14848 provide data with cylindrical and rectangular configurations, varied assembly separation distances, and varied place-ment of the B, C poison.

The critical experiments described in PNL-24388 use stainless steel, Boral, and zircaloy-4 for separation. These materials are used in the Sequoyah spent fuel pool.

The critical experiments described in PNL 2615to use an enrichment of 4.29 wt%

{ 235U (as compared to 2.35 wt% in PNL-2438 and 2.46 wt% in BAW 1484).

The comparison of the KENO modeling with critical experiments is shown in Figure 3. -

The criticals resulted in an average kerr of 0.99273 which represents a KENO method bias of 0.00727.

The standard deviation from the average kerr was calculated to be 0.00404. However,

.when the 'W' testfi for the assumption of normality was applied,it was determined that METilODS 3

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1 the critical kert predictions are not normally distributed. Therefore, the standard devi-ation has been increased by a factor of 1.245 to account for an approximately normal i distribution.'2 This product was further increased by the 95% probability factor (1.645),  !

resulting in an average uncertainty of 0.00827 at the 95% probability /95% confidence level based on the comparison to the critical experiments. This uncedainty was in- I creased further as described below. I The benchmarking results were evaluated for significant trends with respect to enrichment, separation distance, H/U atom ratio, H/235U atom ratio, and poison material i loading. No significant correlations were observed. All variations are well below the '

magnitude of the KENO statistical uncedainty.

This criticality analysis was performed for fuel enriched to 5.0 wt% which is outside the range of enrichment analyzed in the critical experiments. Therefore, a bias was deter-  !

mined conlistent with the method described in the ANSI standard on Nuclear Criticality ,

Safety.13 A linear correlation was developed for the kerr versus hydrogen to 235U atom l ratio in the critical benchmark experiments. This correlation was then used to extrapo- I late to the Sequoyah spent fuel rack value with 5.0 wt% fuel loaded. A bias of 0.00318 Akert was determined and is included in the final kerr evaluation. Also, the KENO method  !

uncertainty was increased by 0.00182 Akertt o allow for the increased uncedainty due to l the enrichment extrapolation.

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CASMO .. . . .

To estimate the uncertainty for the CASMO calculational method used in the determi-nation of equivalent burnup enrichments, a comparison was made between the CASMO and LATTICE'd computer codes. LATTICE is a lattice physics computer code developed by TVA and approved for light water reactor fuel assembly analyses by the Nuclear Regulatory Commission.*d A PWR fuel assembly model was depleted with both CASMO and LATTICE using hot- '

full-power conditions. The reactivity was then calculated at cold conditions and selected exposures out to 20,000 MWD /MTU using the restart features available in both codes.

The maximum difference in the change in reactivity with burnup between LATTICE and -

CASMO was 0.00374 Akerr with the CASMO reactivity change being less and thus more conservative since a lower reactivity change will result in a higher equivalent enrichment.

To increase the conservatism, the maximum difference was then doubled. Therefore, the estimate of the uncedainty in CASMO reactivity due to burnup is 0.00748 Akert.

2.2 COMPUTER CODE APPLICATION KENO was used to establish the required equivalent enrichment at zero burnup that maintains kerr,< 0.95 including uncertainties and biases. CASMO was then used to de-termine the corresponding burnup of 5.0 wt% fuel that has the same reactivity.

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l KENO A KENO model of the spent fuel racks containing fresh, low-enrichment fuel in a heck- i erboard configuration with fresh fuel enriched to 5.0 wt% 235U was analyzed; The enrichment of the low-enrichment fuel was selected such that kerr,5,0.95 when all biases  !

and uncertainties were taken into account.

l CASMO '

The CASMO methodology was used to determine reactivity as a function of exposure for fuel enriched to 5.0 wt%. The steps required are described below.

• I An array of fuel assemblies of 5.0 wt% enrichment was depleted using a lower bound '

on the% vater density (0.60 gm/cm)). This maximizes the reactivity a:Idition at fuel storage conditions due to water density history considerations. The maximum num-ber of burnable absorbers were inserted during the depletion thus maximizing the burnable absorber history reactivity addition when the absorbers are assumed to be removed in the storage racks.

• The reactivity of the depleted fuel assemb!!es in the spent fuel storage racks was de-termined at various burnups. The burnable absorbers were removed from the as-sembly for this calculation. The xenon concentration was set to zero and the remaining isotopes were held constant. <

The reactivity was calculated for fresh assemblies in the spent fuel storage racks at the low-enrichment determined in the KENO calculation and with no burnable absorbers. ,

The reactivity of the fresh low-enrichment assemblies was used to select the burnup of the 5.0 wt% assembly with the same reactivity. The final analysis determined that ,

for an assembly enriched to 5.0 wt% with 20,000 MWD /MTU of accumulated burnup, the equivalent fresh enrichment is 2.71 wt%.

All models used in this analysis used nominal dimensions. The uncertainty resulting  ;

from dimensional tolerances were addressed using sensitivity studies.

2.3 FUEL ASSEMBLY The fuel modeled in this analysis was a fresh Westinghouse VANTAGE SH assembly h without control rods. Burnable poisons were inserted during the CASMO depletions but removed in the spent fuel storage rack calculations, thus accounting for burnable poison history effects which increase reactivity. The uranium in the assembly was conservatively modeled to contain only 235U and 238U at beginning-of-life with no fission product buildup. Depleted assemblies were modeled as previously described using the equivalent enrichment concept. CASMO models the depletion of uranium, production of plutonium and creation of fission products.

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Zircaloy spacer grid straps were modeled. However, the intermediate flow mixer grids were not modeled. Credit was not taken for other assembly structural material such as the top and bottom nozzles. These have been conservatively modeled as water. -

Laterally, the fuel area in both the KENO and CASMO models were divided into a 17X17 array containing the discretely defined fuel rods and guide tubes. The fuel rod cell models the fuel pellet, gap, and clad surrounded by water. The guide tube cell model ,

l contains a water center, the guide tube and surrounding water.

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CASMO is a two-dimensional model and has no axial definition. The KENO model co tains several axiallevels due to the modeling of the assembly spacer grids.

Conservatively, the total active fuel length has been assumed to be the same' enrichment.

Natural uranium axial blankets have not been modeled. Therefore, this model is appil-cable to both the Standard 17x17 and VANTAGE SH fuel assemblies since the fuel rod -

dimenslorn are the same and the zircaloy spacer in the VANTAGE SH fuel assembly is less a neutronic poison than is the inconel spacer in the Standard 17x17 fuel assembly. i 2.4 STORAGE RACK ,

The spent fuel storage rack modeled for this analysis is a high density spent fuel storage rack manufactured by par. The design incorporates the use of the neutron polson ma- '

terial Boral for the purpose of reducing the center-to-center spacing of the storage cells.

The spent fuel storage cell consists of two concentric, square stainless steel tubes, seal welded at the ends. The Boral plate is located in the water-tight void existing between the tubes. These details were explicitly mooeled in both KENO and CASMO.

Axially, the KENO model describes the full Boral plate. All additional rack material has been conservatively modeled as water.

Specific spent fuel storage rack data is provided in Figure 4.

2.5 ANALYTICAL MODELS There are three analytical models used in this analysis. A half-cell model was used for CASMO, while KENO used both a single cell and a multi cell model. A two dimensional CASMO model was used for the burnup analysis. A single-cell, three-dimensional KENO model was used for sensitivity analyses and analyses without burnup. A three-dimensional multi-cell KENO model was used to determine burnup effects using the y equivalent enrichment concept.

CASMO MODEL The half-cell model used by CASMO consisted of a half fuel assembly and the corre-sponding rack and water regions in an infinite radial array. The fuel rods and one cell wall have been explicitly modeled. However, the Boral plate was homogenized with the second cell wall due to CASMO code limitations. Because CASMO is a two-dimensional .

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code, the spacer grids are modeled by homogenizing the grid material with the moder-ator. Appropriate code options were chosen to model an infinite array of such assem-blies.

KENO SINGLE-CELL MODEL .

.The KENO single-cell model consists of one fuel assembly, one storage cell, and the wa-ter region surrounding the storage cell in an infinite lateral array. An infinite array was achieved by choosing the appropriate KENO lateral boundary conditions. The resulting configuration contains only one fuel enrichment, i.

Axially, the single-cell discretely models the fuel assembly spacer grids. Fuel assembly and storage rack material above and below the length of the Boral plate have been conservati(ely modeled as water.

KENO MULTI-CELL MODEL The multi-cell model consists of four quader fuel assemb!Ies which describes fuel of two different enrichments, four quader storage cells, and the water region between the cells, in an infinite lateral array. The infinite array results in a checkerboard configuration containing two different fuel enrichments. As with the single-cell model, an infinite array was achieved by choosing the appropriate KENO lateral boundary conditions.

Axially, the multi-cell model is identical to the single-cell model.

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l 3.0 ANALYSIS RESULTS 4 i

i The reactivity of the spent fuel pool was determined as required by ANSl/ANS 57.2.is The analysis consisted of a combination of CASMO and KENO cases. Uncertainties associ- l ated with pch calculation were included.  !

The CASMO calculations provided the enrichment and burnup relationships required for  ;

this analysis. The KENO calculations used both the single-cell and the multi-cell models. 2 Each KENO resulting in case considers 300 neutron generations and 400 neutrons per generation, j 120,000 neutron histories. The final kerr is the sum of the KENO kerr , the  !

method biases, and the statistical combination of all the uncertainties.  !

3.1 RESULTS i l

Allowing for all biases and uncertainties, the kerr of the spent fuel storage racks has'been determined to be 0.94416 for an infinite array of 5.0 wt% fresh fuel loaded in a checker- i board configuration with 5.0 wt% fuel having accumulated burnup of 20,000 MWD /MTU.

This value was determined as follows:

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kerr = kerr(KENO) + BIASES + UNCERTAINTIES

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The biases include the CASMO and KENO method biases, a boron particle self shielding  ;

allowance, and a bias for the extrapolation of enrichment from the critical benchmark .

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comparisons. The uncerialntles include the KENO statistical uncertainty, the KENO and

.CASMO method uncertaintles, and the mechanical tolerance uncertainty. These values are presenttd in Figure 5 and discussed in the following section.

3.2 SENSITIVITY STUDIES The effect of various parameters on reactivity was determined to ensure the conserva-tism of the analysis. This was accomplished by performing sensitivity studies on these l

parameters with either CASMO or KENO. Due to the statistical variation inherent in the KENO results, cases whose kerr were used directly were run several times with different stariing random number initialization values in an effort to detect any statistical aber-l ration.

L ANALYSIS RESULTS 8

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I WATER TEMPERATURE / DENSITY The KENO criticality analysis was performed at a temperature of 68 degrees F. This was determined to be a conservative value for water temperature and density as follows:

Results from CASMO sensitivity studies show that as the temperature of the water  !

increases and the density of the water decreases, the k- decreases.

The spent fuel storage pool water should always be at or above 68 degrees F because of decay heat from the spent fuel assemblies. Water temperature increases in the  ;

spent fuel storage pool due to decay heat generated by the spent fuel. The Sequoyah FSAR states that the pool water temperature will be maintained at or below 150 de-grees P; CASMO analyses were performed for temperatures of 68 and 212 degrees F and the 68 degree results provided conservative (higher k. ) values. ,

The results of these calculations are shown in Figure 6. t Poison LOADING .

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The criticality analysis was performed with a polson loading of 0.0233 gm/cm2 of 108 in j the Boral. _A total boron content equivalent to or greater than 0.0233 gm/cm2 of 108 was  ;

required by the specification for the Boral used in these racks.18 Since the reactivity de-creases as the poison loading increases,0.0233 gm/cm2 of 108 is the loading assumed.

MECHANICAL TOLERANCES ~ l The effect on the reactivity of mechanical tolerances associated with rack dimensional values has been analyzed. The most significant effect on the reactivity due to mgchanical ,

tolerances is caused by a reduction in the flux trap width. The minimum flux trap width is achieved by combining the maximum cell inner dimension, the maximum sheet metal i thickness, and the minimum cell pitch. This combination of parameters resulted in a ,

change in reactivity of 0.00026 a kerr. This change in reactivity is less than the uncertainty i in the KENO run itself. Therefore, the uncertainty in the KENO run will be used as a l conservative estimate of the reactivity uncertainty due to mechanical tolerances. The uncertalnty associated with the worst mechanical tolerance case calculation is 0.00585 h kert.

PELLET DENSITY l

The analysis was performed using a UO, density greater than can be used at l

Sequoyah.fr This approach gave conservative results when analyzed with CASMO, 1 ASSEMBLY PLACEMENT The analysis was performed with the fuel assembly centered inside the storage cell, since previous criticality analyses 18,18 performed on these racks determined that this was the most reactive configuration.

ANALYSIS RESULTS 9

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CELL DIMENSIONSIBOW This analysis accounts for cell bowing within the analysis for rack cell mechanical toler-ances. The rack cell mechanical tolerance analysis was performed using the worst case combination of maximum cell inner dimension, maximum sheet metal thickness, and minimum cell pitch. Cell bowing causes both the cell pitch and the cell inner dimension to be altered, Therefore, the effects of cell bowing are the same as those addressed in the mechanical tolerance case and do not need to be addressed independently.

AXIAL BURNUP DISTRIBUTION The effects of axial burnup distribution were analyzed using equivalent enrichment data determined by CASMO. The axial burnup distribution was modeled for a bottom peaked, ,

a top peaked, and a cosine shaped distribution. These cases, when compared with an average enrichment case in which no axial variations were modeled, showed a difference of -0.00042 a kerr. The difference is negligible and is ignored. This verifies that the axial distribution of burnup does not affect the criticality analysis.

BORON PARTICLE SELF SHIELDING EFFECT Boron padicle self shielding has been observed in neutron transmission through heter-  !

ogeneous absorber materials. This is a result of neutron streaming between grains of the Boral material. The boron in the Boral was modeled with a homogenous distribution i when actually the boron is in the form of particles distributed throughout the Boral.' This modeling assumption does not account for the particle self shielding.

- 1 Using the neutron transmission value obtained from calculational results,18 the increase l

In kerr due to particle self shielding is 0.003 A kerr. This value was applied to the final kort of the system as a blas. ' '

l BORATED WATER REACTIVITY WORTH An analysis was performed to determine the reactivity worth of the borated water in the

• spent fuel storage pool. Using a 2000 ppm boron concentration, both KENO and CASMO predicted the change in kert to be approximately 17% akerr. This value is not used in the determination of the system kerr but is provided for information.

WATER HOLE REACTIVITY WORTH

[ KENO calculations were performed that show that checkerboarding fresh fuel assemblies with cells filled with unborated water does not increase the reactivity of the system. The kerr for this case is 0.84 which is less than 0.94416 where fresh fuel and burned fuel is considered. Therefore, water holes can replace the burned fuel requirement for imple-menting the results of this analysis.

ANALYSIS RESULTS 10

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1 3.3 REACTIVITY vs ENRICHMENT.

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The sensitivity of the system kert to 235U enrichment was determined using both CASMO and KENO. The results showed that reactivity increases with increasing enrichment, l

' i This relationship is shown in Figure 2.

if a frosh fuel assembly of 5.0 wt% enrichment accuw/ates 6,750 MWD /MTU of burnup, it has an equivalent reactivity to a 4.0 wt% fuel assembly without burnup. Therefore, once a fresh fuel assembly accumulates at least 6,750 MWD /MTU of burnup, the re-l strictions on its placement in the spent fuel storage racks can be removed. This includes l

allowances for the potential effect of burnable absorber history and water density history. ,

Note: The exposure value will be increased by an allowance for monitoring uncerialnty- I as described in the Implementation chapter of this report. '

3.4 ' CREDIT =CR BURNUP Credit for burnup was determined using the equivalent enrichment concept. Fresh fuel assemblies at lower enrichment (2.71 wt%) were used to simulate burned fuel. These depleted assemblies were inserted in a checkerboard storage configuration, along with fuel assemblies at 5.0 wt% representing fresh fuel assemblies. The equivalent enrichment of 2.71 wt% correspnds to a 5.0 wt% fuel assembly depleted to 20,000 MWD /MTU. This infinite checkerboard array of depleted and fresh 5.0 wt% assemblies results in a system kerr,5 0.95.

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The exposure value will be increased by an allowance for monitoring uncertainty as described in the implementation chapter of this report. -

4 3.5 SPECIAL CASES AND POSTULATED ACCIDENTS '

The following discussion addresses credible abnormal occurrences and accident condi-tions for the spent fuel storage pool with respect to criticality. a p DROPPED FUEL ASSEMBLY A dropped fuel assembly accident in the spent fuel storage pool assumes that the as-sembly does not deform the rack structure.20 A fuel assembly dropped in the spent fuel pool can attain one of two configurations; lying on top of the racks forming a 'T' config-( . uration with the rest of the fuel, or between the periphery of the rack and the spent fuel p

J pool wall. These configurations were analyzed and determined to be acceptable as dis-L cussed below:

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• There is a 44.6 cm distance between the top of the active fuel and the top of the -

storage racks (or the dropped fuel assembly). Since the migration length of neutrons in water at 1 gm/cm3 (68 degrees F) is 6.2 cm,21 then the separation distance that ex-1sts can be assumed to be an infinite distance and the stored fuel assemblies do not interact with the dropped assembly.

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Furthermore, for accident conditions, the double contingency principle, ** which re-quires two.unlikely. events to produce a criticality accident, can be applied. It is  !

highly unlikely that a dropped fuel assembly will occur in conjunction with a loss of .

borated water from the spent fuel pool. The wodh of a fuel assembly.on top of the -  :

racks would be much less than the worth of borated water. Therefore, a fuel assem-bly dropped onto the top of the racks would not result in an increase in kerr because -

the wodh of the borated water can be considered.

• - A fuel assembly can come to rest in the space between the racks and the spent fuel pool wall. However, for accident conditions, the double contingency principle,13 can s be applied. Therefore, the analysis for this configuration takes credit for the borated  ;

water normally in the pool.

L An anajysis was performed with KENO that demonstrated that the addition of one fuel assem8ly to the spent fuel rack periphery adds less positive reactivity than the neg-

• ative reactivity associated with the borated water. Therefore, a dropped fuel assem-bly will not result in the kort limit of 0.95 being exceeded.

DROPPED CASK / HEAVY LOAD The Sequoyah Final Safety Analysis Report 22 and Technical Specifications 23 prohibit heavy loads from being moved over fuel assemblies in the spent fuel storage racks. .

There is no effect on the criticality of the spent fuel storage rack.

SEISMIC EVENT The spent fuetstorage racks are designed to withstand loads from a safe shutdown earthquake.22 There is no effect on the criticality of the spent fuel storage rack, o  :

LOSS OF SFSP COOLING A loss of spent fuel pool cooling would result in an increase in the spent fuel pool water. j-temperature. - The effects of increased pool water temperature have been addressed in section 3.2 and are not a criticality concern because reactivity decreases as the water density decreases.

[- MISPLACED FUEL ASSEMBLIES The implementation of this analysis requires administrative controls to be placed on storage of fuel assemblies with enrichment greater than 4.0 wt% and burnup less than 6,750 MWD /MTU. If the fuel is inadvertently misplaced in the spent fuel storage racks ignoring the administrative controls, then the double contingency principle which re-quires two unlikely events to produce a criticality accident can be applied and the pres-

- ence of boron in the spent fuel pool can be assumed.

' ANALYSIS RESULTS 12 c .

W a.

~'

Ik*

nw The worth of the borated water is sufficient to lower the kert of the storage racks to 0.80 s

assuming that the rack is loaded with all fresh 5.0 wt% fuel assemblies. Therefore, in-

' f advertently misloading the fuelin the spent fuel storage racks would not result in a criticality accident.

i 3.6- -

SUMMARY

OF CONSERVATISMS

' Th'ese analyses do not take credit for the following phenomena which would decrease reactivity.

PRESENCli OF BORATED WATER Technical Specification 3.9.1 requires that during refueling operations, the boron con-centration of the refueling canal (and therefore, the spent fuel pool) be 2000 ppm or

-higher. Except for accident scenarios, the presence of borated water and its negative reactivity effect is ignored in this analysis, p.- Y PRESENCE OF BURNABLE ABSORBERS -

c High enriched fresh fuel requires the use of burnable absorbers in order to control power  :

peaking at full power conditions. These burnable absorbers are delivered already in- U

, serted in the fresh fuel assembly. Their effect is to lower the reactivity of the fuel as-sembly' during its first cycle of operation until accumulated burnup effectively lowers the reactivity. _ _ i 1

' ' This analysis assumes that the burnable absorbers are removed from the fuel as,semblies ,.

when placed in the spent fuel storage rack.

i LOWER ENRICHMENT

. 1 a 7

All of these analyses have been completed assuming that all fuel assemblies in the spent fuel racks are enriched to 5.0 wt% All of the assemblies currently in the spent fuel racks  ;

are enriched to less than 4.0 wt% Near term reloads are planned with enrichments less' '

than 4.5 wt% As the enrichment decreases, the reactivity of the fresh fuel assemblies

. decrease. This decreased reactivity due to lower enriched fuel is ignored in this analy-

. sis.

{

, HIGHER BURNUP The reloads currently being discharged from the Sequoyah cores have a region average burnup of 38,000 MWD /MTU. Future reloads are being designed to have discharge burnups of up to 48,000 MWD /MTU. These discharged assemblies, besides having

- enrichments much less than 5.0 wt%, also have burnups considerably greater than the ANALYSIS RESULTS 13 L.

D+

'20 000 MWD /MTU required per this analysis . This increase in burnup and resulting de-

- crease in assembly reactivity is ignored beyond the 20,000 MWD /MTU exposure value.

MISCELLANEOUS CONSERVATIVE ASSUMPTIONS.

Other conservative assumptions in this analysis include:

• - . Ignoring radial neutron leakage from the spent fuel storage racks

• Ignoring the presence of control rods

• -Ignoring the presence of spent burnable absorber assemblies

• Ignoring the higher water temperature of the spent fuel pool. '

• Maximizing burnable poison history effects

• Maximizing water density history effects

• Minimiling the 10B content in the Boral

• Ignoring the presence of axial blankets in VANTAGE SH fuel T

4 e

G ANALYSIS RESULTS 14

.. ~~

.ke ,

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6 4--

4.0 IMPLEMENTATION SCHEDULE AND REQUIREMENTS-The following actions will be taken to implement the results of this analysis.

4 4.1 SCHEDULE Fuel to be delivered in the Summer of 1390 for unit 2 is anticipated *lo have an enrichment greater than 4.0 wt% 235U. Technical Specification and procedure changes must be in place before that fuel is moved from the new fuel vault to the spent fuel storage racks,

[ Note: The new fuel vault at Sequoyah has an enrichment limit of 4.5 wt% 235U.

Furthermore, since there is a unit 1 refueling outage scheduled for the Spring of 1990, "

this opportunity will be used to place the unit 1 discharged fuel in a checkerboard con-figuration in the spent fuel racks. This work will be performed during the unit 1 off load or before fuel with enrichment greater than 4.0 wt% is placed in the spent fuel pool. This minimizes the potential for misloading fresh fuel assemblies face-adjacent to each other when they are moved from the new fuel vault to the spent fuel racks. , ,.

E t

4.2 TECHNICAL SPECIFICATION CHANGES .

Technical Specifications will be modified to require that any fuel assembly with enrichment greater than 4.0 wt% and burnup less than 6,750 MWD /MTU be placed in lo-

' cations in the spent fuel storage racks that have face adjacent cells filled with water or ,

fuel assemblies with at least 20,000 MWD /MTU of burnup.

During fuel movement and until the configuration of the assemblies in the storage racks p is verified to comply with the criticality loading criteria, the spent fuel pool water must be borated with'at least 2000 ppm of boron in order to ensure that a dropped fuel as-sembly or misplaced fuel assembly will not be a criticality concern.

l.TIPLE.TIENTATION SCilEDULE AND REQUIRE.TIENTS 15

p j ~.

4.: j 1

I4.3 FSAR CHANGES Sections 4.3.2.7 and 9.1.2.1 in the Sequoyah Updated Final Safety Analysis Repod must be modified to reflect the new enrichment limit and the administrative controls for placement of fuel in the spent fuel storage racks. '

4.4 -PROCEDURE MODIFICATIONS Sequoyah site procedures will be changed to require that fuel assembly transfer forms be verifled to properly implement the requirements of the proposed technical specifica-tion. The exposure that will be used for the evaluation of whether or not an assembly satisfics itfe burnup criteria will be calculated by multiplying the predicted relative as-sembly. power during the cycle and the measured cumulative core burnup.

The exposure limit against which the assembly exposure is compared includes an un-certainty to account for the uncedainties in assembly power predictions and the core burnup measurement. The uncertainty in the predicted assembly power is 3%.24 The uncertainly in the accumulated core burnup is equal to the uncertainty in the measured core thermal power which is 2%.25 Conservatively adding these two uncertainties and multiplying by two results in a 10% uncertainty to be applied to the assembly exposure criteria. In other words, the assembly exposure limits determined with CASMO will be conservatively increased by 10% (multiplied by 1.10) to allow for uncertainties in the as-

.sembly burnup. For example, the procedures will reflect a criteria of 22,000 MWD /MTU for the assemblies that can be placed adjacent to fresh fuel assemblies when the fresh fuel enrichment exceeds 4.0 wt%.

4.5 PERSONNEL TRAINING l

Fuel handling personnel will be formally trained on the new criticality requirements for the spent fuel storage racks before the first region of fuel enriched to greater than 4.0 wt% is moved into the spent fuel storage pool. .

L 4.6 SPENT FUEL POOL BORATION MONITORING Since the implementation of this analysis requires that borated water be in the spent fuel f

( pool in order to ensure that the criticality margin is maintained in accident scenarios, the boron level in the spent fuel pool must be monitored during any fuel movements. During mode 6, Technical Specification 3.9.1 requires that the boron concentration of the refuel-Ing canal (and therefore, the spent fuel pool) be 2000 ppm or higher. Therefore, site procedures will be modified to ensure that before fuel with enrichment greater than 4.0 wt% is moved, the boron concentration of the spent fuel pool is 2000 ppm or higher.

Every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter, the boron concentration in the spent fuel pool will be assured to be 2000 ppm or higher until fuel movement ceases and the configuration of the fuel in l-IMi'LEMENTATION SCllEDULE AND REQUIREMENTS 16

e. , ,.

k ., : -

j j!; ' M,.; , - 4;. '

s 3

t i

the spent fuel storage racks has been verified to correctly implement the criticality load-

.ing criteria.  ;

.-4.

b t

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z IAIPLE31ENTATION SCilEDULE AND REQUIRE.T1ENTS 17 i

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1 5.0

SUMMARY

AND CONCLUSIONS By administratively controlling the placement of high reactivity fuel in the spent fuel storage racks at Sequoyah, the enrichment limit can be increased from 4.0 wt% to 5.0 wt% whilegaintaining required criticality safety margins.

Fuel enriched to greater than 4.0 wt% with burnup less than 7,500 MWD /MTU (6,750 MWD /MTU increased by 10% to allow for burnup uncertaintles) shall be placed only in-Sequoyah spent fuel storage rack locations with face adjacent cells which contain:

Fuel assemblies with accumulated burnup of at least 22,000 MWD /MTU, or

• . Water.

The spent fuel pool water shall contain at least 2000 ppm of boron during fuel movement and until the configuration of the fuel in the spent fuel storage racks has been verified to correctly implement the criticality loading criteria. .

l

.o l

l

'. SU.\lMARY AND CONCLUSIONS 18

~ ~ ~ "

ry

~ ~

3 m' ' '

., 4 'j

%,3 < 3.M.

1

~

,7 Le l i

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4

, 6.0 ' REFERENCES H

~

1. :Sequoyah Updated Final Safety Analysis Report, Chapter 4.1. I 2.' L WCAP,10444-P-A, Addendum 2-A, " VANTAGE SH Fuel Assembly," Aprili 1988. 'l

' 3. 'NUREG/CR-0200;" SCALE: A Modular Code System for Performing Standardized j Compiater Analyses for Licensing Evaluation,"(SCALE 3.1 version). ,

i 4. . CAM-009-UG, "CASMO-3:-A Fuel Assembly Burnup Program - User"s Manual," Re- t c

'visioni1, Malte Edenius, Ake Ahlin,.Bengt H. Forssen, Studsvik/NFA-86/7.

3

5. ."CASMO-3: Benchmark Against Yankee Rowe isotopics," Peter J. Rashid,

~

y

. Studsvik/SOA-86/05,' September,' 1986, Studsvik/NFA-86/7. '

6. "CASMO-3 Benchmark Against Critical Experiments," Per Jernberg, August,1986, Studsvik/NFA 86/11. ~

7. . " Determining'Burnup Credit Requirements-in Spent-Fuel Storage Racks Using Reac- '

{

U

, . . tivity Equivalencing," W.A. Boyd, ANS Transactions,1987 Winter Meeting, Volume 55.

1. 8l B AW-1484-7, M. N, Baldwin, G. S. Hoovier, R. L. Eng, F. G. Welfare, " Critical Exper-

'i iments Supporting Close Proximity Water Storage of Power Reactor Fuel," July 1979.

9. PNL-2438, S. R. Bierman, B. M. Durst, E. D. Clayton, " Critical Separation Between  :

L Subcritical Clusters of 2.35 wi% 235U Enriched UO Rods in Water With Fixed Neutron

10. P L-26 5, S B erm , B. M. Durst, E. D. Clayton, " Critical Separallon Between Suberitical Clusters of 4.29 wt% 235U Enriched UO Rods in Water With Fixed' Neutron -

~ Poisons," NUREG/CR-0073, May 1978.

11r ANSI N15.15-1974," Assessment of the Assumption'of Normality." .

.12. Probabillity and Statistics for Engineers , Irwin Miller, and John Freund, Second Edi-tion, Prentice-Hall,' inc.,1977, page 231. -

I <

'13. ANSI /ANS-8.1-1983," Nuclear Criticality Safety in Operations With Fissionable Mate--

rials Outside Reactors."

14. TVA-TR78-02A dated April,1978," Methods for the Lattice Physics Analysis of LWRs," '

' approved.by NRC October 16,1979.

.15. ANSI /ANS-57.2-1983," Design Requirements For Light Water Reactor Spent Fuel-  !

Storage' Facilities at Nuclear Power Plants." '

16. par Systems Corporation Quality Control Procedure, QCP-82-9001, " Specification for

~

i Neutron Absorber Plates Sequoyah Nuclear Power Plant Units 1 and 2," Revision 3, December 12,1979, s 17. Letter from Westinghouse to David Marks (TVA), dated February 15,1989, on the

subject of fuel pellet density (No.JPS-89-011).

18. Westinghouse Summary of Criticality Analysis for 4.0 wt% 235U , as documented in TVA memo LOO 811013228 and Westinghouse letter FP-TV-471 dated December 29, 1981.

f

[ REFERENCES 19 jil

[1 , ,

. , l ', *

). . ;p . U',.

f. ,

'e

.19. " Summary Report - Nuclear Criticality Analysis for the Spent Fuel Pool of the Sequoyah Nuclear Power Plant of the Tennessee Valley Authority," Revision 1, Nu -

clear Associates International (NAI), May 2,1980, transmitted as pad of par Design Report DR-9001-1.

20. Sequoyah Updated Final Safety Analysis Report, Chapter 9.1.2.

21. Nuclear Reactor Engineering , Glasstone, Samuel, Sesonske, Alexander, Van Nostrand Reinhold Company,1967, page 147.

22. Sequoyah Updated Final Safety Analysis Report, Chapter 9.1.2.3.

23. Sequoyah Unit _1 and Unit 2 Technical Specifications, Section 3.9.7.

24. Sequoyah Updated Final Safety Analysis Report, Chapter.4.3.3.3.

25. Sequoyah Updated Final Safety Analysis Report, Chapter 15.1.2.2.

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REFERENCES 20

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. 7.0 ILLUSTRATIONS -

ti. 9:-

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-, ILLUSTRATIONS ~

21 4 +

i 31 I;r'..,.. ' ' ,

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--FIGURE 1 CASMO K-INFINITYL VERSUS BURNUP '

1-5.0 W/0 -

68 DEG F b'

t: 0.9 - -

PoR RACKS Z 17X17 YANTAGE 5H ~

C. .

z .. 1 i

0.6 -

I

-O 2

tn. ,l j 0.7 - ,

ti .i i

- 0.6

(

, , , , , 1

.0 10000 -20000 ,.

30000 40000 50000 60000 f

-BURNUP (MWD /MTU) -

~.;

FIGURE 2 -

o-CASMO K-INFINITY VERSUS ENRICHMENT 1- .

L l l .

!: ZERO BURNUP #

h.

68 DEG F

. Z - 0.95 - PcR RACKS  !

'E

E 1 -

M .

O h 0i90-

. O .'

y . .

0.85

,. 3 4 5

ENRICHMENT (W/O U-235) r:

1 1

L. l v.

l PAGE 22

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[

Characterizing KENO Reference Test Poison Separation (cm) K-eff 1-Sigma B AW-1484 - 01 None 0.000 0.99250 0.00268 BAW 1484 02 None 0.000 0.98677 0.00224 BAW-1484 03 None 1.636- 0.99593 0.00210 4

B AW-1484 04 B4C-Pins 1.636 0.98779 0.00278 7 B AW-1484 -05 B4C-Pins 3.272 0.98254 0.00273 BAW-1484 -06 B4C-Pins 3.272 0.99006 0.00388 B AW-1484 07 B4C-Pins 4.908 'O.98460 0.00281 B AW-1484 08 B4C-Pins 4.908 0.99077. 0.00356 BAW-S84 09 None 6.544 0.98657 0.00317 PNL-2438 15 None' ' 11.920 0.99647 0.00238 J PNL-2438 05 None 8.390 0.99288 0.00282 PNL-2438 22 None 6.390 0.99138 0.00284

- PN L-2438 - 21 None 4.460 0.99078 0.00231 PNL-2438 34 Stainless 10.440 0.99950 0.00266 PNL-2438 35 Stainless 11.470. 0.99822 0.00244 PNL-2438 26- Stainless 7.760 0.99005 0.00276 PNL-2438 27 Stainless ~ 7.420. 0.99469 0.00248 PNL-2438 20 Boral 6.340 0.99944 0.00278 PNL-2438 -16 Boral 9.030 0.99157 0.00278 PNL-2438 46 Zirc-4 8.790 0.99658 0.00248 PNL-2438 47 .Zirc-4 8.780 0.99217 0.00294 Pt1L-2615 30 Zirc-4 10.920 0.99499 0.00299

" -PNL-2615 .29 Zirc-4 10.860 -0.99618 0.00285-PNL-2615 31 . Boral 6.720 0.99674

• 0.00321 PNL-2615 14 Stainless 8.580 0.99141 0.00302 PNL-2615 13 Stainless 9.650 0.99369 0.00290 PNL-2615 08 Stainless 9.220 0.99646 0.00313 PNL-2615 07 Stainless 9.760 0.99304 'O.00303 PNL-2615 04 Noi.e 10.640 0.99449 1 0.00347 PNL-2615 06 Aluminum 10.720 0.99612 0.00290 PNL-2615 05 Aluminum 10.770 0.99134 0.00297 PNL-2615 10 SS 1.05%B 6.100 0.98918 0.00341 PNL-2615 09 SS 1.05%B 8.080 0.99590 0.00302 PNL-2615 12 SS 1.62%B 5.760 0.99459 0.00320 PNL-2615 11 SS 1.62%B 7.900 0.99018 0.00281 35 points Average 0.99273 +/-0.00404 Figure 3. Results of KENO Benchmarking: Comparison of KENO With Critical Experiments ILLUSTRATIONS

_ 23 ji _ _ _ _ _ _ _ - - - - - - - - - - -

e J

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' j l

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ACTUAL- ANALYZED i RACK DESCRIPTION '

. . ~1 Rack Cell Array.

Cell Pitch, in.

42 by 33 Infinite j 10.375 10.375' I RACK CANISTERS Material 304 Stainless Steel- 304 Stainless Steel- 1

. . Inner Can inside Dimension,'in. 8.75- 8.75 lyer Can Thickness, in.. 0.090 {

0.090 Inner Can Outside Dimension,In. 8.93 8.93

• Outer Can Inslde Dimension, in.

9.33 9.33 Outer Can Thickness, in. 0.0360 0.0360 Outer Can Outside Dimension, in. 9.402 9.402 o

NEUTRON' POISON -

Material Boral Boral -4

^

. Total Length, in. -.

~

147 147 ~"

.. Total Thickness, in. ~ 0.10 0.10 Width, in. 8.625 8.625 Sheath Material- Aluminum - Aluminum Sheath Thickness, in. 0.01 0.01 Core Material B4C-Al B4C-Al  ;

Core Thickness, in. 0.08 0.08 108. Density, gm/sq cm 0.0233(minimum) -0.0233, t

Figure 4. par spent Fuel storage Rack Data: Cornparison of Actual vs Analyzed l

l l '.

4 I

l- l l

L ILLUSTRATIONS .1 24 '

, u l

p; 3;j: a

'. h ?. '

q$+

y ,qs-i sf x

i 1' 7 o

BASE K-EFFECTIVE (KENO) 0.91617

' BI ASES (AK-EFFECTIVE) '

' KENO METHOD 0.00727

- ENRICHMENT

g. EXTRAPOLATION ALLOWANCE 0.00318 BORON PARTICLE SELF-SHIELDING 0.00300 TOTAL BIAS 0.01345 *

'UNCERTA NTIES(AK-EFFECTIVE) -

STATISTICS OF KENO RUN 0.00441 KENO METHOD AND ENRICHMENT EXTRAPOLATION 0.01009

' MECHANICAL TOLERANCE 0.00585  ?

CASMO BURNUP 0.00748 ** -

' ~'

TOTAL UNCERTAINTY 0.01454 *

(SQUARED, SUMMED, ROOTED)

+

l FINAI K-EFFECTIVE DETERMINATION Kerr = Ken (KENO) + BIASES + UNCERTAINTIES 1

1

'0.94416 = 0.91617 + 0.01345 + 0.01454 Figure 5. Final K effective Determination: S wt% Fuelin Checkerboard Configuration l

l

.1

, r

.. i ILLUSTRATIONS 25

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fof 9:

L

+ -

FIGURE 6 .

REACTIVITY vs WATER DENSITY

[ --

,e '1

.(

l'

.i. '

O.95- '

i

{

t O PPM BORON 0.90- PoR RACKS

! -t

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r 0.85-Z :l'

C- 0,80- '

- .Z

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-h 0.75-g 1- <

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0.70- '"

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0.65- t.

L 0.60- ,

L 0.55- a i .

h, 0 0.2 0.4 0.6 0.8 1 . i WATER DENSITY (gm/ cms)

F L

1 I:

PAGE 26 g 6

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, . _ . .